1. Field Of The Invention
The present invention relates to nuclear fuel rods for a nuclear reactor fuel assembly and in particular to a cladding tube made of a zirconium alloy and lined on its inner surface with a zirconium liner.
2. Description Of Prior Art
A nuclear fuel element typically has a cladding tube which houses a stack of fuel pellets of sintered uranium oxide, thorium oxide, plutonium oxide or a mixture of such oxide fuels, and end plugs which seal both the upper and lower ends of the cladding tube. In addition, a gas storage plenum or space is provided within the cladding tube to accommodate fission product gases. Within the plenum, a spring or mechanism is provided for stably holding the sintered fuel pellets within the cladding tube.
The cladding functions in part to prevent contact and to prevent chemical reaction between the fuel pellets and the coolant and/or moderator. In addition, the cladding tube prevents contamination of the coolant by radioactive fission products emitted from the fuel pellets resulting from reactor operation. The loss of leak tightness of the cladding tube could contaminate the reactor and associated systems and interfere with plant operations.
The material for the cladding tube is required to have excellent mechanical properties and high corrosion resistance in the environment and under the operating conditions expected during reactor operations. Common cladding materials include zirconium and its alloys and stainless steel Zirconium alloys in which the major component is zirconium are widely used as the materials of the cladding tube. Two of the most commonly used zirconium alloys are Zircaloy 2 and Zircaloy 4 and are described in ASTM standard B350-91 (1991), Standard Specification For Zirconium and Zirconium Alloy Ingots For Nuclear Application, compositions R60802 and R60804, respectively Zircaloy 2 (composition R60802)is composed of from 1.20 to 1.70 weight percent tin, 0.07 to 0.20 weight percent iron, 0.05 to 0.15 weight percent chromium, 0.03 to 0.08 weight percent nickel, where the iron plus chromium plus nickel content is 0.18 to 0.38 weight percent, and the balance is zirconium plus impurities. Zircaloy 4 (composition R60804) is composed of from 1.20 to 1.70 weight percent tin, 0.18 to 0.24 weight percent iron, 0.07 to 0.13 weight percent chromium, where the iron plus chromium content is from 0.28 to 0.37 weight percent, and the balance is zirconium plus impurities. The maximum impurity levels for Zircaloy 2 and Zircaloy 4 are given in the following table which is from Table 1 of the ASTM B350-91 Standard.
______________________________________ MAXIMUM IMPURITIES, WEIGHT % R60001 R60802 R60804 ______________________________________ Aluminum 0.0075 0.0075 0.0075 Boron 0.00005 0.00005 0.00005 Cadmium 0.00005 0.00005 0.00005 Carbon 0.027 0.027 0.027 Chromium 0.020 -- -- Cobalt 0.0020 0.0020 0.0020 Copper 0.0050 0.0050 0.0050 Hafnium 0.010 0.010 0.010 Iron 0.150 -- -- Hydrogen 0.0025 0.0025 0.0025 Oxygen * * * Magnesium 0.0020 0.0020 0.0020 Manganese 0.0050 0.0050 0.0050 Molybdenum 0.0050 0.0050 0.0050 Nickel 0.0070 -- 0.0070 Niobium 0.010 0.010 0.010 Nitrogen 0.0065 0.0065 0.0065 Silicon 0.0120 0.012 0.0120 Tin 0.0050 -- -- Titanium 0.0050 0.0050 0.0050 Tungsten 0.010 0.010 0.010 Uranium (total) 0.00035 0.00035 0.00035 ______________________________________ *When so specified in a purchase order, oxygen shall be determined and reported. Maximum or minimum permissible values, or both, shall be as specified in the purchase order.
Although several zirconium alloys such as Zircaloy 2 and Zircaloy 4 have excellent properties as a cladding material under normal conditions, at high degrees of burn up, the cladding tube is subject to brittle fracture due to localized mechanical stresses resulting from differential expansion and friction between the fuel and the cladding combined with the presence of fission product gases released from the fuel but trapped within the cladding. This type of failure mechanism is referred to as Pellet Cladding Interaction (PCI).
In order to prevent PCI cracking, a cladding liner material is positioned between the fuel pellets and the zircaloy tube. This liner eliminates direct contact between the zircaloy and the fission products. In order to relieve the local stresses caused by the thermal expansion of the fuel pellets, the liner material must also have a high ductility. It is well known that a suitable material for the liner is pure zirconium. The maximum impurity levels for nuclear grade zirconium are listed in the preceding table under the heading R60001. Both crystal bar and sponge zirconium have been used for this purpose. Since the zirconium needs to be very ductile, it performs this function best if it is unalloyed. However, unalloyed zirconium has very poor corrosion resistance. If the cladding is breached and water and/or steam enters the interior of the fuel rod, the zirconium liner, and particularly the inner wall of the zirconium liner is subject to rapid oxidation which may cause abnormally large secondary failures. Such failures can lead to large axial splits in the cladding and allow unacceptably large releases of radioactive material into the coolant. However, adding alloying material to the liner to improve the corrosion resistance can also reduce the ductility of the liner.
The ductility and the large grain size of the zirconium liner may promote surface cracking or microfissuring during the fabrication of the cladding. U.S. Pat. No. 4,894,203 provided an improved nuclear fuel rod where a layer of an alloy of zirconium having additions of iron, chromium, copper, nitrogen or niobium which form the alloy is placed on the inner wall of the zirconium liner in order to inhibit the tendency of the zirconium liner to crack during fabrication and oxidize during operation. U.S. Pat. Nos. 4,610,842 and 4,816,215 discuss alloying the entire liner with between 0.1 and 4% tin to improve its corrosion resistance. However, because substitutional alloying by tin increases the strength of zirconium-tin alloys beyond that of pure zirconium, the ductility of liners made of zirconium-tin alloys is decreased.
It would thus be an advantage over the prior art to provide a zirconium liner for cladding of nuclear fuel rods having a corrosion resistant inner alloy surface while maintaining the ductility of the zirconium liner.